ML20054L146

Duke Power Co Nuclear Station Post-Accident Liquid Sampling Sys Sampling & Control Panels Functional Testing Rept,Nov 1981

ML20054L146
Person / Time
Site:	Catawba
Issue date:	11/30/1981
From:	Orth W DUKE POWER CO.
To:
Shared Package
ML20054L140	List: ML20054L139 ML20054L146
References
NUDOCS 8207070189
Download: ML20054L146 (59)

Text

_.

G 2 8I.9 DUKE POWER COMPANY NUCLEAR STATION POST ACCIDENT LIQUID SAMPLING SYSTEM Sampling and Control Panels Functional Testing Report November 1981 r

by: Wm. C. Orth

'.;,w Staff Chemist Power Chemistry Scientific Services Steam Production Dept.

8207070189 820628 PDR ADOCK 05000413 A

PDR

ABSTRACT This report gives the findings and results of the functional testing of the Duke Power Company designed and developed Post Accident Liquid Sampling and Control Panels. Conductivity and pH,-sample dilution, and gas stripping were tested with simulated samples. The final results given are within the limits of good analytical practice and meet design criteria.

O

e INTRODUCTION Timely information on the characteristics of the reactor coolant and contain-ment atmosphere can be of great use to nuclear power station personnel during accident conditions.

Reactor coolant analyses can provide information on the level of core damage and on the potential for chemically induced equipment r

degradation.

Containment atmosphere analyses can provide additional informa-tion on core damage and reactor coolant system integrity, as well as provide an early indication on the potential if any for significant offsite doses.

Most nuclear power stations routinely obtain information concerning the radio-logical and chemical composition of fluids in various systems through a nuclear sampling system.

However, following an accident, significant amounts of radioactivity may be released from the core, causing abnormally high radiation levels throughout the station.

As a result of these higher radiation levels, sampling systems that perform adequately during normal station operation may be of little use during accident conditions.

Problems could arise from a lack of radiation shielding causing the sampling area to be inaccessible, from the sampling system not being designed to provide the' sample needed, or from a host of other considerations that may not have been included in the original sampling system design.

In any event, the necessity of having to use only the normal sampling systems could, and at TMI-2 did, result in significant delays in obtaining information that would aid station personnel in diagnosing and mitigating accident conditions. To prevent these delays, a nuclear power station design should provide the capability either to sample in the presence of abnormally high radiation levels, or to obtain the same information by a means other than sampling.

i The Duke Power Company Post Accident Liquid Sampling System is described in Duke Design Manual #NSAC/23.

Prototypes of the liquid sampling panel and the control panel were installed in the Research Laboratory at the Training and Technology Center for testing analytical functions. The tests and results are given in this report.

n.,..-

PURPOSE The pre-installation testing was done to determine that the Post Accident Liquid Sampling Panel will adequately perform the functions for which it was designed (Table 1).

These functions include:

1.

Isolating and cooling a portion of liquid sample.

2.

Depressurizing and degassing the liquid sample.

3.

Measuring conductivity and pH of the sample.

4.

Taking an increment of the sample and making a proportional dilution.

5.

Diluting the gas stripped from the sample to a known volume at atmospheric pressure.

6.

Rinsing and draining all portions of the panel except the isolated diluted liquid and gas samples.

All of these functions of the sampling panel are operated from a remote control panel.

i

O e

Table 1 Post Accident Sampling System Design Criteria 1

1.

Sampling Conditions

- liquid:

temperature 70 F - 650 F pressure 0 psig - 2500 psig

- containment air:

temperature 70 F - 300 F pressure 0 psig - 60 psig

- reduce sample conditions to:

temperature

< 200 F pressure s O psig 2.

Radiation

- limit personnel exposure to less than 5 rem whole body and 75 rem extremities.

- obtain samples with radiation levels of l0 Ci/cc for liquids and 0.1 Ci/cc for containment atmosphere.

- for grab samples, reduce sample dose rate by a factor of 1000.

3.

Sample Analysis

- perform sample analysis within following ranges:

i boron concentration

> 100 ppm dissolved gases 10-2000 cc/kg 0 STP pH 3-14 conductivity 0-10" umhos radiation:

liquid 1pCi/cc - 10 Ci/cc containment atmosphere 1pCi/cc - 0.1 Ci/cc 4.

Station Interfaces

- keep system compact, but allow for maintenance accessibility.

- minimize need for station auxiliaries.

- must operate in following environment:

temperature 70 F - 100 F relative humidity 20% - 90%

pressure 0 psig 5.

Human Factors

- system should be simple to operate under stressful conditions.

I i

FUNCTIONAL TEST PROCEDURE The preliminary design procedure given in EPRI Report NSAC/23 Ja. 1981 prepared by Duke Power Steam Production Department was used to begin the test.

In the process of following this procedure, design changes and modifications were found necessary. A stepwise final operating procedure is given in Appendix B to this report.

Samples to test the pH and conductivity analysis and liquid sample dilution functions were prepared in stock solutions.

They entered the sampling panel by gravity feed from an elevated container.

No high pressure or temperatures were introduced in this phase of the testing.

The stock solutions were analyzed with laboratory instruments after being prepared as standards.

A red dye was used as the stock solution in the dilution testing.

This provided a direct colorimetric analysis by absorption to eliminate analytical error as much as possible.

The red dye solution also provided visual evidence as to the adequacy of flushing and draining after sampling.

The time required to flush to eliminate all evidence of pink color in the outlet water could be observed easily.

The degree of dilution of the liquid sample in the design is intended to minimize the exposure to an analyst when collecting and analyzing the grab samples.

Grab samples can be collected in a syringe in amounts of one to two milliliters or less of a sample that has been diluted as much as 6000 to 1.

This would mean a postulated post accident condition of 10 Ci/g would be reduced to 1.7 millicuries i

in a milliliter of grab sample collected which could be handled safely in the labora tory.

The diluted grab sample will be used for a gamma isotopic analysis and boron analysis.

The same sample can be used for counting and chemical analysis or two samples can be taken.

A Parr reaction vessel was used to prepare dissolved g'as samples.

These samples were introduced to the sampling panel at elevated temperature and pressures.

The 2000 ml capacity reaction vessel was filled to overflowing with demineralized Then 500 ml of water was displaced by hydrogen gas from a cylinder of water.

I 1

k _.

compressed gas.

The heating element was turned on until the water was heated to the desired temperature while the stirrer operated.

Pressure increased from the expansion of the gas and the steam pressure. Additional hydrogen pressure was added to give the desired sample pressure. The hot pressurized sample was admitted-to the sampling panel. The sample was saturated with hydrogen for the given temperature and pressure.

Steam pressure from the tables was subtracted from the gage pressure to obtain the partial pressure of hydrogen.

In the sampling panel the hydrogen was stripped from the water sample and diluted with 1000 ml of nitrogen (where the analysis for nitrogen is also desired, argon gas can be used for stripping and diluting). The dilution is accomplished by fillin'g an evacuated vessel until the pressure is atmospheric. A portion of the diluted gas is drawn off with a syringe.

The syringe can be used for isotopic analysis and the contents introduced into a gas chromatograph for gas analysis.

I l

RESULTS pH and CONDUCTIVITY METERING The liquid sampling panel has builtin conductivity and pH probes with meters and readout on the control panel. The isolated sample, after degassing, is let down.into the chambers where the probes are located. This operation is conducted remotely from the control panel.

Fol. lowing the panel operating procedure Appendix B, the pH meter was stan-dardized at pH 7.41 0 25 C using a solution of Fisher Certified B-82 Buffer Salt, dry (Potassium Phosphate Monobasic -

Sodium Phosphate Dibasic).

Filtered water from the laboratory tap was used as a sample into the panel using the operating procedure.

Readings were made for pH and conductivity with laboratory instruments on the tap water.

Results were as follows:

Lab Meters Panel Meters pH 6.6 Cond. 0.10 x 103 umhos pH 6.6 Cond. 0.09 x 103 pmhos The pH meter was restandardized at. 9.18 0 25 C using a solution of Fisher Certified B-80 Buffer Salt, dry (Sodium Tetraborate.)

A neutralized boric acid sample solution was prepared and run thru the sample panel using the operating procedure.

Results were as follows:

Lab Meters Panel Meters 3

3 pH 9.12 Cand. 1.68 x 10 umhos pH 9.2 Cond. 1.62 x 10 umhos The amount of sample going to the pH and conductivity cells was sufficient to provide accurate readings.

LIQUID SAMPLE DILUTION A small portion of the collected sample that has been let down into the pH probe chamber is isolated in a short tubing loop. This increment is blown into a sample cylinder from which it is flushed with demineralized water and air into the 3000 ml dilute sample vessel.

Ratios of 1:1000 and 1:1500 were chosen but there was some variation in the actual dilution volume measured.

Start up calibration of each panel is neces-sary to determine and eliminate dilution volume errors. The concentration of the diluted sample is determined and from that the concentration of the original

sample collected can be calculated:

diluted ppm x ml dilution volume = original concentration ppm 1

0.55 ml Volumes were measured for sample increment that is isolated by the 3-way valve (ADV). The valve was activated 20 times for one sample and 10 times for another.

The water was collected and weighed.

20 increments = 12.0871 g = 0.60 ml/ increment 10 increments = 5.0886 g = 0.50 ml/ increment r

An average value of 0.55 ml is used in subequent calculations.

(This is a correction of the value of 1.2 ml given in the original design manual).

The i

water flowmeter was checked by measuring the volume of water delivered to the dilute sample cylinder. The calibration of the flowmeter had to be adjusted to measure 500 ml 25 ml.

In order to test the accuracy of the liquid sample dilution by the sampling panel the operating procedure was followed using only one increment of the ADV with a 5 second bypass flow. The total volume of diluted sample was collected in a graduated cylinder to substantiate the volume. A portion of the diluted sample was analyzed.

Results are given in Table 2.

A solution of Bordeux Red dye was used as a sample entering the panel using a standard value of 1000 ppm.

Analysis was done by spectrophotometry using a Bausch & Lomb Spectronic 88 at 590 nm wave length and 10 cm path.

Table 2 i

Diluted Sample, ppm Concentration, ppm Dilution Ratio Actual Volume by Analysis for Original Sample

=

1:1000 0.5 5/500 ml 1.14 1036 1:1000 0.5 5/475 ml 1.22 1053 1

1:1000 0.5 5/500 ml 1.10 1000 1:1000 0.5 5/500 ml 1.10 1000 1:1000 0.5 5/490 ml 1.20 1069 1:1000 0.5 5/485 ml 1.25 1062 1:1500 0.5 5/750 ml 0.72 982 1:1500 0.5 5/740 ml 0.73 982 1:1500 0.5 5/740 ml 0.72 969 1:1500 0.5 5/740 ml 0.73 982 1:1500 0.5 5/745 ml 0.71 962 1:1500 0.5 5/720 ml 0.83 1086 j

blank 500 ml 0.08 73 i

Concentration known Value = 1000 ppm; Experimental mean = 1015 t 43 ppm.

DISSOLVED GAS The liquid sampling panel isolates 150 ml of the sample under operating pressure.

The sample may be precooled by use of a cooling coil For high concentration of hydrogen >750 cc/kg, the sample needs to be isolated hot and allowed to cool to

<250 in the sample vessel before degassing.

Degassing and dilution are accomplished simultaneously. The dilute gas sample vessel is evacuated to <25"Hg.

When the pressure returns to atmospheric the total volume of stripped gas and dilution gas is equal to the volume of the vessel, 1000 ml.

The volume of the liquid sample vessel is 150 ml. Therefore the concentration of hydrogen in the original sample is calculated:

%H2 in dilute sample x 10 = cc/kg 150 To simulate samples containing dissolved hydrogen, a Parr Reactor stainless steel vessel No. 4522, 2000 ml capacity, was used. This reactor vessel allows a water sample to be heated and pressurized with the addition of hydrogen gas while stirring. The operating limits for the reactor are 1400 psig and 350 C.

Tests run at pressures from 100 to 1150 psig and temperatures from 200 F to 500 F incorporate hydrogen saturation concentrations from 77 cc/kg to 2047 cc/kg.

The reactor vessel was filled with filtered water (2037 ml).

500 ml of the water was displaced with hydrogen gas and the reactor was heated while stirring. When the desired temperature was reached the hydrogen gas pressure was applied to reach the desired pressure for the test.

~

The pressure in the vessel was used to inject the sample into the sampling panel using the sampling operating procedure.

The control panel pressure gage confirmed the pressure on the reaction vessel.

Samples of diluted gas (N2 dilution) were removed with a syringe and needle thru a septum in the gas grab sampler.

The samples were analyzed with a gas chromatograph using a thermal conductivity detector. The results are l

shown in Table 3.

The results are tabulated for three temperatures and the saturation constant shown obtained from the data S= cc/kg. This is PSIA 4 compared with a calculated value using the formula S=8.45 x 10 P/H given in " Water Coolant Technology of Power Reactor" Paul Cohen pg. 107 with Henry's Law Constant.

PSIA was obtained by subtracting steam tables pressure from the gage pressure read.

A graph presentation is given in Fig. 1.

i

c i

• Table 3 Temp H2 Part Press cc/Kg @ STP Deviation F

PSIA Calculated PASP Published PASP-Publish 200 87 108 116 100

+ 16 187 232 212 220 8

287 356 232 340

-108 387 480 459 500

- 41 487 604 462 570

-108 587 728 529 710

-181 S = cc/kg S = 1.24 5=1.05 5 = 1.19 PSIA 300 33 57 71 50

+ 21 133 228 148 130

+ 18 233 400 270 280

- 10 333 572 522 400

+122 433 744 546 520

+ 26 533 915 739 700

+ 39 S = cc g S = 1.55 5 = 1.43-5 = 1.23 500 200 871 842 910

- 68 300 1306 989 1250

-261 470 2047 1810 1800

+ 10 S = cc k S = 4.36 5 = 3.79 5 - 4.18 3 = 69 4

l 1

I l

t l

c

c Figure 1 2'000 _

- -- - c-CalculatedLHenr_y'~s Law).

/.

_. : 2.. ;= n = = = : ----

f, 1 : = L

_L J

..D....____-_,_.___,

__g Published (Seidel etc.)

J, '.. _-

.1

.=_"..U._'=.".__=8,_2_2=..i

- :/ c]

A PASP (Duke 1982)----

r-

-- :z

n_ _. c = t _ _ : e a :== ;-- r4

_u

2

_L.._

• -...._=_i.__=__=_.__.....,.___=_.

4

,..f._.---

1800 i

s

• '* l.

7

. ::_.:1':

-- - - g 2

r u.._.

-. =-*!:'... ; _... ; _ s m t

~

2.m - 3 =:22: 2:: = -

r 22 = - r,(._i ;. k _a_,'=_=__.. '-... _, _. :.

.i.,.. ~. : 1 - ^ - _ : : = -- ~ - - - -._ _._ J~. ____. _,

=_..p._r =.__: = 1-t

_ n t_--

~

_4 e m. _ t.. o ra. ---- :. =n _: :==;===;.=r=;- w-g

.3

- ! -- y g / r : --

: r :._.nt = n_ _, : - ul :- 21_ _

f, j ~ 4 1600 -

- o i- -!

u :- 3 n.I : g....: 1. -- _ ;- 7 ---] = u2 = t-- t. 2.2 2:=

__1 -

p___._ _...._ _.._ ____,

4

---L---

..

. i :.: ~ _. :_. :)a.g.. __.,._....__ + :==..f:._.=_ = s_.

3___..

7

_. _ _3

__ j

, re' _ir F;il_'E Hc---1 EE=},,_

i 4

__ _ _ _j. --+ /.

, _ f: ; _t=:22 m :=a12:= r,-

nu; x = ;= :

1 l

rn= :f r zf =j = rrh-- --~1rt w _

r_

r

.=.r._- } = :=t- -, c==e

~

1400 3

.,i -.

,. ; p n _ 7= J_:=_ ;.. _._l _..- _ _,~_-. __ _- - -'.

2.

. i f.

. ; p _- ~r. !.. ; = n:r= u

=1_c.--,-

q

-Jr Tr- : = f rr.-t.rt=r__1_

.4 4

..f.

.r

..c

. j_. :. a_,._...

_.__{ =._r.... _.._ _,

f-

- t.

i

~'I 1200

/

• j.

1

. F.

. r. _ l ? : % 1721. '"l_

~_Z '"*

21 : = * = t==

~

r-

.7

_'..;_. 1

... _. _L

-i-- :

- r A=_=_. j:

_ _.a

't

.- i

.1

?

J /*J !_-

f-4

F. _,:. _ h_

q

.e t

p

= t a0 '

L=l, =_ = >

f _

t t

~

[

1..1 x t. -- _. :{". J 9@: *- z

/

i 4

...a. a

= t =2.. _a

. :y

- - - - - - - l-i 7

ll i

-t

t

'. c t

. $ q.. _..__,t.:: =

w.

a a

=.

r i :. r:. =

1000

=:=u- : - -

n - - - - n - - - =~1

/.

A y

a[

-- r--..

+.

t:.

.c at==;===:

a

.z != t. ;--*

p.. ;

.t.

q

. 4

.._.7..._.

1 y: I : - Ti-

= E =N. E = f-~=- Ai y

. :. =

= q=r,<= :

.t i

/

f

=_:1--*xg f

I.

"- - - [ c

.i _:._~._. t

_ ' ^

3

^

g 3I _ _ ___

.L $

.;]

800 _

,f 4

l

.S < & --

1

.,. _. - -._.l.

!__ _ =

t

--n--

. ir

..p r.p_r..A = j =_.

c. q ;=_ ;__ :._ =_. f _

l.

-,j.

o i

[

^

- I

'{- --*

. _.. _. _. _../... __ __ __..s..

~- -

. - -EE- - - :

e

[

~

-l -

[;

l-'-e s

5 - --

_,3 - b- -~=g ; t - -

600

-t=

=

1 o

i i

.g *!

J - i-, _ j f L. li=rcl= ~ -~j

!. _ _. _ _ _ _ _ _ _ l

_ _.... _ _ _ _ _.j/

. i.+ :

q_._-

. Z : h ! -* = ZI. :--d

.l T.

i -

- [-

A 1_

l A

t l

1

-*-_!=-4.=

_.7 _-

i j

l 1l g!f.

L1_1 L.. c,

I E..= l 2_~~"ir~ 'r

---J i

2 J

.l-

-_4 400 l i

g F p 3. _. y - - -y". r. ~ n t

= ~--~

p

_ : :.: inn = = ---i

+ ::

x

.c L

i ~

e 7. g.=___g=, _ ___3 1

i. _.

l.,

,I: _". } r_.n. _ 1 =_. r---j J.

.f

1..

.-o C_,_

9nn

._y_-

. _. _ _4...l _ __

j j

l

.i i.l'"E'j[-j=-3 4

I.

-I

__. -.... 1 r_:_; =. _1 GG i

6 8

a - - - -

/-

i r

p 1=-

c ;rm==:

I~~

~' - l t

..._4

..:n

=.__:

j i

n

{ --i _

_ _ _. j. -

p....

{._ _.

2.. _

_.. _J.._._ _.j__..___j..._ __h..= _h 1.

_r 2= a l

+

t

.t r=

=.. ; m:.: =1 i

100 200Paan u. Pc'tk u ph00 R00 600 a

m --

CONCLUSIONS Test results show that the design criteria have been met:

1.

pH and conductivity readings are equivalent to laboratory bench instru-ment results.

2.

Samples can be diluted in the panel within 43 ppm of the selected concentrations.

3.

Dissolved gas can be stripped from saturated samples of water.

Deter-I minations were made for hydrogen in filtered water which approached the theoretical values, calculated from Henry's constant. The results are comparable to other published test results in the Battelle Memorial Institute Topical Report, Seidell BMI T-25 5/18/80.

1 4.

The sampling panel can be flushed and drained of all sample liquid by an operator remotely, with the control panel. Only the isolated diluted portion of liquid and stripped gas are present when the operator approaches the sampling panel for a grab sample.

P i

l l

l l

i

\\

l i

r e

RECOMMENDATIONS The liquid dilution factor used will depend on the post-accident radiation level in the sample water.

Dilutions of 1000:1 or greater will limit the accuracy of boron determinations.

Presently colorimetric methods that use small portions of sample (1 ml) are preferred for minimal exposure to the analyst. Research is needed for a better method of boron analysis in the ppb range.

All dissolved hydrogen saturation data, both theoretical and published observations, are for water solutions.

Research is needed to show whether these values are valid for boric acid (2000 ppm B) solutions or if not, data and tables should be prepared. The apparatus used to test this panel and the panel procedure can be used for this research.

After the liquid sampling panel is used with actual radioactive coolant samples the need and means for decontamination should be developed.

i i

I l

[

1.

.i APPENDIX A PREOPERATION CALIBRATION 1.0 LIQUID DILUTION The water'from the water flow meter should be collected and the volume measured. Adjustment should be made to the meter and flow regulation valve until consistant 20 ml can be obtained. Tests should be run at 500 ml and 1000 mi settings.

l 2.0 TEMPERATURE j

With no sample to panel, the temperature should be ambient at each j

temperature setting.

3.0.pH & CONDUCTIVITY Calibration is part of the Preparation step in the Operating Procedure.

The conductivity meter is not adjustable by the operator.

I 4.0 FLUSH AND DRAIN The time required for each step in the Flush and Drain cycles will depend somewhat on local conditions of the panel installation.

A run-thru of the cycles should be performed while observing the time required ~ until water runs freely from the sample return line during Flush or stops flowing into the panel sump during Drain.

s 1

I l

1 I

~

t

APPENDIX B OPERATING PROCEDURE 1.0 PRE - PREPARATION 1.1 Check all connections and valving of sample lines, and supply lines to the sampling panel. Sample lines include containment sump and reactor coolant.

Supply lines includes air (80 to 100 psi), inert gas (argon or nitrogen) 40 psi, demineralized water, and cooling water.

110 VAC also required.

1.2 The pH buffer solution should be prepared for a buffer around 7 pH.

Stock buffer mix may be used and at least one liter prepared. The conductivity of the buffer solution should be measured by a reliable conductivity meter outside of the panel and recorded.

1.3 Turn selection knob to reset position (Fig.1).

1.4 Manual valves for sample and supply air, gas, and water to the panel should be opened.

1.5 Turn System Power key to "on" position right.

2.0 PREPARATION 2.1 Turn selection knob to Panel Prep position 1.

2.2 Press Selection Power activate button.

2.3 Press Purge button 18 and hold 20 secs, and release.

2.4 Press Calibrate button 1A and hold until conductivity and pH meters (Fig. 2) stabilize, then release.

2.5 Record conductivity noting any discrepancy from conductivity of standard.

Adjust pH meter to standardize at buffer value.

2.6 Press Purge button IB and hold 20 secs, and then release.

2.7 Press Flush button IC and hold until conductivity and pH meters stabilizc, then release.

2.8 Press Purge button 1B, hold 20 seconds, then release.

I 2.9 Press Drain button 10, hold 2 seconds, then release.

3.0 SAMPLE CIRCULATION 3.1 Turn selection knob to Sample Recirc position 2.

3.2 Set temperature meter to position Tc 1.

3.3 Press Selection Power activate button.

L

3.4 When Tc 1 temperature stabilizes turn selection knob to Sample position 3.

4.0 SAMPLE ISOLATION i

l 4.1 Set temperature meter to position Tc 2.

4.2 Press Selection Power activate button.

4.3 When Tc 2 temperature stabilizes record temp and press button 3A, stop sample.

4.4 When pressure on panel meter stabilizes record pressure and press button 38, isolate sample.

4.5 Turn selection knob to Depressurization position 4.

5.0 DEPRESSURIZATION & GAS STRIPPING 5.1 Reset gas flow meter to zero. Toggle switch should be to the right, preset counter at 99999.

5.2 Press selection Power activate button.

5.3 Start gas flow with button at the meter (Fig. 2).

5.4 Watch level meter which should be reading 25 - as soon as needle begins moving up, quickly press Gas Flow stop button.

Read gas flow meter and record volume.

5.5 If level does not stop in + range, press Increase button 4A.

5.6 Turn selection knob to Liquid Sample oosition 5.

6.0 SAMPLE ANALYSIS 6.1 Press Selection Power activate button.

6.2 Press Conductivity button SA and hold until cond. meter stabilizes.

Record reading.

6.3 Press pH button 5B and hold until pH meter stabilizes.

Record reading.

6.4 Press Gas Sample button SC and hold 1 second.

6.5 Press Diluted Gas Grab Sample button 5F.

6.6 Turn selection knob to Liquid Sample Prep position 6.

7.0 SAMPLE DILUTION 7.1 Press Selection Power activate button.

7.2 Press Sample Increment button 6A and allow 5 seconds to operate.

7.3 Reset water flow meter for desired volume of dilution water, and press start button (Fig. 2).

Let water flow until volume selected is complete.

7.4 Press Mix button 6B and hold 10 seconds.

7.5 Turn selection knob to Liquid Sample position 7.

8.0 DILUTE GRAB SAMPLE 8.1 Press Selection Power activate button.

8.2 Press Dilute Sample Flow button 7A.

8.3 Press Dilute Grab Sample button 7B.

8.4 Turn selection knob to Flush position 8.

9.0 PANEL FLUSH 9.1 Indicator light should be on.

If light is not on press Reset button.

9.2 Press activate button 8A once for light to go out.

9.3 Press activate button for 1st cycle and wait minutes.

9.4 Press activate button 8A for 2nd cycle and watch pH and conductivity meters until they drop to minimum readings and stabilize.

9.5 Press activate button 8A for 3rd cycle, wait 3 minutes.

9.6 Press activate button 8A 4 times until " complete" light comes on.

9.7 Turn selection knob to Drain position 9.

10.0 PANEL DRAIN 10.1 Press Selection Power activate button.

10.2 Press activate button 9A once to turn out " complete" light.

10.3 Press activate button 9A again for 1st cycle wait minutes.

10.4 Press activate button 9B again for 2nd cycle wait minutes.

10.5 Press activate button 9B again for 3rd cycle wait minutes.

10.6 Press activate button 4 times until " complete" light comes on.

10.7 Return selection knob to Reset position.

10.8 Turn System Power key to left to operate sump pump.

Pump stops at low i

level.

11.0 DECONTAMINATION 11.1 If radiation reading is high, sump may be flushed by turning selection knob to Panel Prep position 1.

11.2 Turn System Power key to the right and press Selection Power activation button.

11.3 Press and hold Flush button 1C for 2 minutes.

11.4 Return selection knob to Reset and turn System Power key to the left to start sump pump. Repeat step 1-4 as often as is necessary.

Note:

Time increments in 9.3, 10.3, 10.4, 10.5, are determined at each installation during calibrations, i

t

APPENDIX B Figure 1 I

DRAIN.

OFF RESET PANEL PREP.

Ok Q,C{"f Q Q

O O.0-

+i

. c /lA< A. Arje Mes/ A;.

[

e-

.s

. ~.

FLUSH, 9.

1.!

. SAMPLE RECIBC.

N 0

. $ lce$n

~

s-ooo 8

2_ -

.,.b

}(no

-- c LIQUID SAMPLE ~

7-3-

SAMPLE

'IA& Flav S to f 3 A &

'O 7 & G e<>$

fsol0ft3B Q g

LIQUID SAMPLE PREP.

DEPR ESSURIZATION.

IQUID SAMPL c&

TA Q c d.

Q o

'A @ s.,-pa ke O p 1,,c, a,,, eJ GAS SAMPLk l

cc Q now

\\

l

.L.

rs @ Iurns 9A Q ir,ereaso q-ffhjef(!*st 4*A & afeevenst a @ ni, r,@ c,nl

~

l 1

,. cc.

SYSTE,'.t POWER i SELECT 10N POWEH,,

- 4.W:.

5.- '

u, :..~

~..,

g

'8'.

..t;;. _'-

v@ Q

=

Q_

=-

SEE DR AWING NO. LO401800 FOR PANEL DETAIL jr! ~J'

/th.

[C VO I VO 0

o e

APPENDIX B Figure 2 LOW H6GH O

O E G E '^^ as ALARM V

-CTA RT STCP REOET a

,a-

.Lincnoubos/cu -

t,..,......, D..,L WATE R..*

.3 y

-,2.

s'.

. uE Asuns i.,7

.A..

~

.1..

\\

-..se e

.e,

. OFF CH6 G.

1 1

o CK 1

r.

.w l

e l

1 l

[P

~

j p

• 1

~

l w

O O CD S ~ CD '

r 1

. ~

_. k.". * '-

~ ~ '

• LOW HlCH

.e

- pr. nr sTo, nesty-o

.o-e 1

.0 0

ALARM

.N pH D1 O'

bd MEASURE

' 314 l

3000

',r.s I-CFF bHECK

+

T

+

O

=

.+.

1.e.

+

-s.

.s

• E

.s,. f.

OF Ed3 n

M COOLANT SYSTEM'

-~,..

4 f

pf(,B LOG. NO R. SUMP:

O. SAMPLE INLET v.2-

.O.SAuPi.s ouTur a

O o. s.... wArea A ADl ATION - 4 4

s

~

[*-

\\ -

MONITO R t~

,Tc G

w, L

3 COUNTEM

~

^

4-

,,1.e,,..','.

y' a-s

-- f, *, '.- ' %, ~,

' '. 'q:;.-

y-

++;

. ; e :.

~,..

e....

s,

..s....

. =*.-

-s. -m.

,3

.r.,,--

,p..

-*... -,7.. -,;-

,.v..:

,w,.".

y].i.

~ ~...-; ;, ;. _,.

. r.c.i.

-n r-

-. : s;..,

a.

.e..

r r.

I

~

APPENDIX B F

Figure 3-1.

b*

- - em.

_.ar:remr - suaa:we.=.ar j

TOIN,fmNAL M ::=-

vg t Spaay hl ADE R 34 D

'J

{

06t uf f D Gas r1

.J C21954w.L(4 I

i

,f 10 04 5

' ggY I

y L

CLN I

F 4 pH

'O 4 pH L. l-SOLUTION

$0tullON

[

-)

l l

l n

a a

1C3

!., L a M

j E

SAWMt RifuRN +--J E

lg=

r 7 sccg,,,i!.., ;,,..Wgf ri v-l rm eqp

~

x j eg p-4 p p

=

i

m nm X

~M,.h.y

,.i.

5 ll 8

n' il l

p.

ll

{

es 11

)

I

.I 3

d' M

• M f

' I DitUf f 0 ts0Us0 '

[

~-

(

I N22: F

~O 8I "#LI r*

i[

c g g @id, a

.,w o

\\

c=y,7onna o

m i

.....,[h,b'[ _7j k

[

i i

c fi *m N..N,.......

g

- s>u

{i pr-a >i C

{,-j non uinas EQ

.c'. ?.*'.. @$ p !

ii

~ '

_ Ml M ;-.Aa O~l 3

7.Q'f.,M

$ W J lil l

l.

- r v

- c h-Qn I

a ai

!.mm.......L.c y,T_..g;--[.")

Yo. p, f.

yo-Ns.,4f E A SU.M

.Nt.

c e

....e L.

Sim.ttINLET -b F

=,,1

+--- Ot

+

i i;

i C C n

r

=

su

\\

o s

I I

i I-

~

CNS 410.9 Expand and clarify your discussion of the reactor coolant pressure (5.2.5) boundary (RCPB) leakage detection systems.

Describe how your systems meet each of the positions of Regulatory Guide 1.45.

Provide infor-mation to show how total identified leakage flowrate is measured.

Indicate the method to be used to obtain an accuracy of 1 gpm or better in one hour for unidentified leakage by the cyclic operation of sump pumps.

Describe more fully the use of the volume control tank level for monitoring RCPB leakage flowrate.

For each source of intersystem leakage describe the three separate detection methods as recommended in Regulatory Guide 1.45, position C3.

Response

See revised Section 5.2.5.

410.10 Provide the specific K values determined in your criticality eff (9" I* 1) analysis for the new fuel storage arrangement with the associated assumptions and input parameters.

Clarify your assumption regard-ing water moderation when maximizing K Also, verify that the eff.

new fuel storage racks are capable of maintaining a K,ff of 0.98 or less under optimum moderation (foam, small droplets, spray or fogging) or identify the means provided for preventing such a con-dition in the new fuel storage vault.

Describe the seismic and tornado qualification of the railroad freight door into the new fuel storage building and the features inside the building to prevent tornado missiles, entering through a damaged or missing freight door from reaching safety related equip-ment.

Response

The criticality analysis of the new fuel storage racks is discussed in Section 9.1.1.3.

The railroad freight door was not designed for tornado missile impact.

There are no safety related systems or components that can be impacted if the missile were to penetrate the freight door.

410-5 Rev. 6 i

t CNS R commendation GL-5 The licensee should upgrade the AFW system automatic initiation signals and cir-cuits to meet safety grade requirements.

Rasponse Sem response to Recommendation GS-7.

(4) Response to the questions identified in Enclosure 2 of the M' arch 10, 1980, letter follows:

Question 1 a.

Identify the plant transient and accident conditions considered in establish-ing AFWS flow requirements, including the following events:

s i

1)

Loss of Main Feed (LMFW) 2)

LMFW w/ loss of offsite AC power 3)

LMFW w/ loss of onsite and offsite AC power 4)

Plant cooldown 5)

Turbine trip wi h and without bypass j

6)

Main steam isolation' valve closure 7)

Main feed ifne break 8)

Main steam line break

/

9)

Small break LOCA 10)

Other transient or accident conditions not listed above.

b.

Describe the plant protection acceptance criteria and corresponding technical bases used for each initiating event identified above.

The acceptance criteria should address plant limits such as:

1)

Maximum RCS pressure (PORV or safety valve actuation) 2)

Fuel temperature or damage limits (DNB, PCT, niax'imum f0el central tem-f perature) 3)

RCS ' cooling rate limit to avoid excessh;e~ coo 1 nt shrinkage 3

4)

Minimum steam generator 1N el to assure sufficient steam generator heat f

transfer surface to remove decay heat and/or cool down the primary system s

<g

/

j

,s 4

R2sponse 4a

~

..m x.

1. a The Auxiliary Feedwater System serves aca backup system for supplying feed-water to the secondary ~ side of the, steam generators at times when the "eed-water system is not available, thereb/ maintaining the heat sink capabillt,ies of the steam generator.

As an thgineered Safeguards" System, tbs. Auxiliary Feedwater System is directly reli 1 upon to prevent core damagV6nd sy" tem T

3 overpressurization in the event of transients such as a loss of normalife'ed-

~

water or a secondary system pipe rupture, and to provide a peajs# for plant cooldown following any plant transient.

6

/

s.. Z.

./

J l

^

410-21 s

Rev. 6

~'

CNS Following a reactor trip, decay heat is dissipated by evaporating water in the steam generators and venting the generated steam either to the condensers threugh the steara dump or to the atmosphere through the steam generator safety valves or the power-operated relief valves.

Steam generator water inventory r.ust be maintained at a level sufficient to ensure adequate heat transfer and continuation of the decay heat removal process.

The water level is maintained under these circumstances by the Feedwater System, or if the Feedwater System is not operable, by the Auxiliary Feedwater System which delivers an emergency water supply to the steam generators.

The Auxiliary Feedwater System must be capable of functioning for extended periods, allowing time either to re-store normal feedwater flow or to proceed with an orderly cooldown of the plant to conditions where the Residual Heat Removal System can be placed into operation for continued decay heat removal.

The Auxiliary Feedwater System flow and the emergency water supply capacity are sufficient to remove core decay heat, reactor coolant pump heat, and sensible heat during the plant cooldown.

The Auxiliary Feedwater System can also be used to maintain the steam generator water levels above the tubes following a LOCA.

In the latter I

function, the water head in the steam generators serves as a barrier to pre-vent leakage of fission products from the Reactor Coolant System into the secondary plant.

DESIGN CONDITIONS The reactor plant conditions which impose safety-related performance require-ments on the design of the Auxiliary Feedwater System are as follows for the Catawba Units.

Ii-Loss of Main-Feedwater Transient Loss of main feedwater with offsite power available f

Station blackout (i.e., loss of main feedwater without offsite power available) l 3

Secondary System Pipe Ruptures Feedline rupture Steamline rupture Loss of all'AC Power Loss of' Coolant Accident (LOCA)

Cooldown

' Loss of Main feedwater Transients The design loss of main feedwater transients are those caused by:

laterruptions of the Main Feedwater System flow due to a malfunction in the feedwater or condensate system 410-22 Rev. 6

CNS Loss of offsite power or blackout with the consequential shutdown of the system pumps, auxiliaries, and controls Loss of main feedwater transients are characterized by a reduction in steam generator water levels which results in a reactor trip, a turbine trip, and auxiliary feedwater actuation by the protection system logic.

Following reactor trip from a high initial power level, the power quickly falls to decay heat levels.

The water levels continue to decrease, progressively un-covering the steam generator tubes as decay heat is transferred and discharged in the form of steam either through the steam dump valves to the condenser or through the steam generator safety or power-operated relief valves to the atmosphere.

The reactor coolant temperature increases as the residual heat in excess of that dissipated through the steam generators is absorbed.

With increased temperature, the volume of reactor coolant expands and begins fill-ing the pressurizer.

Without the addition of sufficient auxiliary feedwater, further expansion will result in water being discharged through the pressur-izer safety and/or relief valves.

If the temperature rise and the resulting volumetric expansion of the primary coolant are permitted to continue, then (1) pressurizer safety valve capacities may be exceeded causing overpressur-ization of the Reactor Coolant System and/or (2) the continuing loss of fluid from the primary coolant system may result in bulk boiling in the Reactor Coolant System and eventually in core uncovering, loss of natural circulation, and core damage.

If such a situation were ever to occur, the Emergency Core Cooling System would be ineffectual because the primary coolant system pres-sure exceeds the shutoff head of the safety injection pumps, the nitrogen overpressure in the accumulator tanks, and the design pressure of the Residual Heat Removal Loop.

Hence, the timely introduction of sufficient auxiliary feedwater is necessary to arrest the decrease in the steam generator water levels, to reserve the rise io reactor coolant temperature, to prevent the pressurizer from filling to a water solid condition, and eventually to estab-lish stable hot standby conditions.

Subsequently, a decision may be made to proceed with plant cooldown if the problem cannot be satisfactorily corrected.

The blackout transient differs from a simple loss of main feedwater in that emergency power sources must be relied upon to operate vital equipment.

The loss of power to the electric driven condenser circulating water pumps re-sults in a loss of condenser vacuum and condenser dump valves.

Hence, steam formed by decay heat is relieved through the steam generator safety valves or the power-operated relief valves.

The calculated transient is similar for both the loss of main feedwater and the blackout, except that reactor coolant pump heat input is not a consideration in the blackout transient following loss of power to the reactor coolant pump bus.

Secondary System Pipe Ruptures The feedwater line rupture accident is postulated to result in the loss of feedwater flow to the steam generators but also results in the complete blow-down of one steam generator within a short time if the rupture should occur downstream of the last nonreturn valve in the main or auxiliary feedwater pip-ing to an individual steam generator.

Another significant result of a feed-line rupture may be the spilling of auxiliary feedwater out the break as a 410-23 Rev. 6

CNS consequence of the fact that the auxiliary feedwater branch line may be con-nected to the main feedwater line the region of the postulated break.

Such situations can result in the injection of a disproportionately large fraction of the total auxiliary feedwater flow (the system preferentially pumps water to the lowest pressure region) to the faulted loop rather than to the effective steam generators which are at relatively high pressure.

The system design allows for terminating, limiting, or minimizing that fraction of auxiliary feedwater flow which is delivered to a faulted loop or spilled through a break in order to ensure that sufficient flow is delivered to the remaining effective steam generator (s).

The concerns are similar for the main feedwater line rup-ture as those explained for the loss of main feedwater transients.

Main steamline rupture accident conditions are characterized initially by plant cooldown and, for breaks inside containment, by increasing containment pressure and temperature.

Auxiliary feedwater is not needed during the early phase of the transient but flow to the faulted loop will contribute to an ex-cessive release of mass and energy to containment.

Thus, steamline rupture conditions establish the upper limit on auxiliary feedwater flow delivered to a faulted loop.

Eventually, however, the Reactor Coolant System will heat up again and auxiliary feedwater flow will be required to be delivered to the non-faulted loops, but at somewhat lower rates than for the loss of feedwater transients described previously.

Provisions are made in the design of the Auxiliary Feedwater System to limit, control, or terminate the auxiliary feed-water flow to the faulted loop as necessary in order to prevent containment overpressurization following a steamline break inside containment, and to en-sure the minimum flow to the remaining unfaulted loops.

Loss of All AC Power The loss of all AC power is postulated as resulting from accident conditions wherein not only onsite and offsite AC power is lost but also AC emergency power is lost as an assumed common mode failure.

Battery power for operation of protection circuits is assumed available.

The impact on the Auxiliary Feed-water System is the necessity for providing both an auxiliary feedwater pump power and control source which are not dependent on AC power and which are capable of maintaining the plant at hot shutdown until AC power is restored.

Loss-of-Coolant Accident (LOCA)

(

The loss of coolant accidents do not impose on the auxiliary feedwater system any flow requirements in addition to those required by the other accidents l

addressed in this response.

The following description of the small LOCA is provided here for the sake of completeness to explain the role of the auxiliary l

feedwater system in this transient.

I Small LOCA's are characterized by relatively slow rates of decrease in reactor coolant system pressure and liquid volume.

The principal contribution from the Auxiliary Feedwater System following such small LOCAs is basically the same as the system's function during hot shutdown or following spurious safety injection signal which trips the reactor.

Maintaining a water level inventory l

410-24 Rev. 6 New Page

CNS in the secondary side of the steam generators provides a heat sink for re-moving decay heat and establishes the capability for providing a buoyancy head for natural circulation.

The auxiliary feedwater system may be utilized to assist in a system cooldown and depressurization following a small LOCA while bringing the reactor to a cold shutdown condition.

Cooldown The cooldown function performed by the Auxiliary Feedwater System is a partial one since the reactor coolant system is reduced from normal zero load tem-peratures to a hot leg temperature of approximately 350 F.

The latter is the maximum temperature recommended for placing the Residual Heat Removal System (RHRS) into service.

The RHR system completes the cooldown to cold shutdown conditions.

Cooldown may be required following expected transients, following an accident such as a main feedline break, or during a normal cooldown prior to refuel-ing or performing reactor plant maintenance.

If the reactor is tripped fol-lowing extended operation at rated power level, the AFWS is capable of deliver-ing sufficient AFW to remove decay heat and reactor coolant pump (RCP) heat following reactor trip while maintaining the steam generator (SG) water level.

Following transients or accidents, the recommended cooldown rate is consistent with expected needs and at the same time does not impose additional require-ments on the capacities of the auxiliary feedwater pumps, considering a single failure.

In any event, the process consists of being able to dissipate plant sensible heat in addition to the decay heat produced by the reactor core.

1. b Table Q410.33-1 summarizes the criteria which are the general design bases for each event, discussed in the response to Question 1.a above.

Specific assumptions used in the analyses to verify that the design bases are met are discussed in response to Question 2.

The primary function of the Auxiliary Feedwater System is to provide suffi-cient heat removal capability for heatup following reactor trip and to remove the decay heat generated by the core and prevent system overpressurization.

Other plant protection systems are designed to meet short term or pre-trip fuel failure criteria.

The effects of excessive coolant shrinkage are eval-uated by the analysis of the rupture of a main steam pipe transient.

The maximum flow requirements determined by other bases are incorporated into this analysis, resulting in no additional flow requirements.

410-25 Rev. 6 New Page

CNS Qu;stion 2 Dsscribe the analyses and assumptions and corresponding technical justification ussd with plant condition considered in 1.a above incluaing:

a.

Maximum reactor power (including instrument error allowance) at the time of the initiating transient or accident.

b.

Time delay from initiating event to reactor trip.

c.

Plant parameter (s) which initiates AFWS to flow and time delay between initi-ating event and introduction of AFWS flow into steam generator (s).

d.

Minimum steam generator water level when initiating event' occurs.

e.

Initial steam generator water inventory and depletion rate before and after AFWS flow commences -- identify reactor decay heat rate used.

f.

Maximum pressure at which steam is released from steam generator (s) and against which the AFW pump must develop sufficient head.

g.

Minimum number of steam generators that must receive AFW flow; e.g.,

1 out of 2? 2 out of 4?

h.

RC flow condition -- continued operation of RC pumps or natural circulation.

i.

Maximum AFW inlet temperature.

J.

Following a postulated steam or feed line break, time delay assumed to iso-late break and direct AFW flow to intact steam generator (s).

AFW pump flow capacity allowance to accommodate the time delay and maintain minimum steam generator water level.

Also identify credit taken for primary system heat removal due to blowdown.

k.

Volume and maximum temperature of water in main feed lines between steam generator (s) and AFWS connection to main feed line.

1.

Operating condition of steam generator normal blowdown following initiating event.

n.

Primary and secondary system water and metal sensible heat used for cooldown and AFW flow sizing.

n.

Time at hot standby and time to cooldown RCS to RHR system cut in temperature to size AFW water source inventory.

Response

Analyses have been performed for the limiting transients which define the AFWS parformance requirements.

These analyses have been provided for review and have b :n approved in the Applicant's FSAR.

Specifically, they include:

410-26 Rev. 6 i

t W r d % m

CNS Loss of Main Feedwater (Station Blackout)

Rupture of a Main Feedwater Pipe Rupture of a Main Steam Pipe Inside Containment In addition to the above analyses, calculations have been performed specifically for Catawba Units to determine the plant cooldown flow (storage capacity) require-m;nts.

The Loss of All AC Power is evaluated via a comparison to the transient results of a Blackout, assuming an available auxiliary pump having a diverse (non-AC) power supply.

The LOCA analysis, as discussed in response 1.b, incorpo-rates the system flows requirements as defined by other transients, and therefore is not performed for the purpose of specifying AFWS flow requirements.

Each of the analyses listed above are explained in further detail in the following sec-tions of this response.

Loss of Main Feedwater (Blackout)

A loss of feedwater, assuming a loss of power to the reactor coolant pumps, is des-cribed in FSAR Section 15.2.6.

It is shown that for a station blackout transient the peak RCS pressure remains below the criterion for Condition II transients and no fuel failures occur (refer to Table Q410.33-1).

Table Q410.33-2 summarizes the assumptions used in this analysis.

The analysis assumes that the plant is initially operating at 102% (calorimetric error) of the Engineered Safeguards Design (ESD) rating shown on the table, a very conservative assumption in defining decay heat and stored energy in the RCS.

The reactor is assumed to be tripped on low-low stsam generator water level, allowing for level uncertainty.

As shown in the FSAR, there is a considerable margin with respect to filling the pressurizer for a loss of normal feedwater transient with or without power to the reactor coolant pumps.

Rupture of Main Feedwater Pipe Tha double ended rupture of a main feedwater pipe downstream of the main feedwater lina check valve is analyzed in FSAR, Section 15.2.8.

Table Q410.33-2 summarizes the assumptions used in this analysis.

Reactor trip is assumed to occur when the faulted generator is at the low-low level setpoint (adjusted for errors).

This conservative assumption maximizes the stored heat prior to reactor trip and mini-mizes the ability of the steam generator to remove heat from the RCS following re-actor trip due to a conservatively small total steam generator inventory.

As in the loss of normal feedwater analysis, the initial power rating was assumed to be 102% of the ESD rating.

Auxiliary feedwater flow of 492 gpm was assumed to be dnlivered to the 2 non-faulted steam generators 1 minute after reactor trip.

The criteria listed in Table Q410.33-1.

This analysis established requirements for layout to preclude indefinite loss of auxiliary feedwater to the postulated break, and establishes train association re-quirements for equipment so that the AFWS can deliver the minimum flow required in 1 minute assuming the worst single failure.

410-27 Rev. 6 New Page

CNS Rupture of Main Steam Pipe Inside Containment B cause the steamline break transient is a cooldown, the AFWS is not needed to re-move heat in the short term.

Furthermore, addition of excessive auxiliary feed-water to the faulted steam generator will affect the peak containment pressure following a steamline break inside containment.

This transient is performed at I

four power levels for several break sizes.

Auxiliary feedwater is assumed to be l

initiated at the time of the break, independent of system actuation signals.

The maximum flow is used for this analysis.

Table Q410.33-2 summarizes the assump-tions used in this analysis.

At 30 minutes after the break, it is assumed that the operator has isolated the AFWS from the faulted steam generator which subse-quantly blows down to ambient pressure.

The criteria stated in Table Q410.33-1 are met.

This transient establishes the maximum allowable auxiliary feedwater flow rate to a single faulted steam generator assuming all pumps operating, and establishes lay-out requirments so that the flow requirements may be met considering the worst sin-gle failure.

Plant Cooldown Maximum and minime, flow requirements from the previously discussed transients most the flow requirements of plant cooldown.

This operation, however, defines the basis for minimum required condensate storage tank level, based on the re-quired cooldown duration, maximum decay heat input and maximum stored heat in the I

system.

As previously discussed in response 1.a, the auxiliary feedwater system l

partially cools the system to the point where the RHRS may complete the cooldown, i.e., 350 F in the RCS.

Table Q410.33-2 shows the assumptions used to determine l

the cooldown heat capacity of the auxiliary feedwater system.

The cooldown is assumed to commence at the maximum rated power, and maximum trip dalays and decay heat source terms are assumed when the reactor is tripped.

Pri-mary metal, primary water, secondary system metal and secondary system water are all included in the stored heat to be removed by the AfWS.

See Table Q410.33-3 for the items constituting the sensible heat stored in the NSSS.

This operation is analyzed to establish minimum tank size requirements for aux-iliary feedwater fluid source which are normally aligned.

410-28 Rev. 6 New Page

CNS Qusstion 3 Vorify that the AFW pumps in your plant will supply the necessary flow to the steam gInerator(s) as determined by items 1 and 2 above considering a single failure.

Id2ntify the margin in sizing the pump flow to allow for pump recirculation flow, scal leakage and pump wear.

Rasponse a)

The AFW pumps will supply the necessary flow to the steam generators consider-ing a single failure.

b)

The Catawba AFW pumps are provided minimum flow protection by automatic re-circulation control valves.

These valves do not recirculate any flow during normal operation of the AFW pumps unless the flow rate through the pumps ap-proaches the minimum flow value.

This is not a continuous recirculation sys-l tem.

No margin is allowed for pump recirculation flow.

c)

A margin of 10 gpm is available to account for system leakage.

d)

A margin of 3% is available to account for pump wear.

l l

410-29 Rev. 6 Carry Over

TABLE Q410.33-1 CRITERIA FOR AUXILIARY FEEDWATER SYSTEM DESIGN BASIS CONDITIONS Condition or Additional Design Transient Classification

• Criteria

• Criteria Loss of Main Feedwater Condition II Peak RCS pressure not to exceed design pressure.

No consequential fuel failures.

Station Blackout Condition II (Same as LMFW)

Pressurizer does not become water solid.

Feedline Rupture Condition IV 10 CFR 100 dose limits.

Core does not uncover.

RCS design pressure not exceeded.

Loss of all A/C Power N/A Note 1 Same as blackout assuming turbine driven pump.

Loss of Coolant Condition III 10 CFR 100 dose limits 10 CFR 50 PCT limits Condition IV 10 CFR 100 dose limits 10 CFR 50 PCT limits Cooldown N/A 100 F/hr 557*F to 350 F

• REF:

ANSI N18.2 (This information provided for those transients performed in the FSAR).

NOTE 1:

Although this transient establishes the basis for AFW pump powered by a diverse power source, this is not evaluated relative to typical criteria since multiple failures must be assumed to postulate this transient.

410-30 Rev. 6 New Page m

+

e r --

e

TABLE Q410.33-2 (Page 1)

SUMMARY

OF ASSUMPTIONS USED IN AFWS DESIGN VERIFICATION ANALYSES L'oss of Feedwater Main Steamline Break Transient (Station Blackout)

Cooldown Main Feedline Break (containment) a.

Max reactor power 102% of ESD rating 3651 MWt 102% of ESD rating 0, 30 102% of rated (102% of 3581 MWt)

(102% of 3581 MWt)

(percent of 3425 MWt) b.

Time delay from 50 sec 2 sec 21 sec variable event to Rx trip c.

AFWS actuation 10-10 SG level NA 10-10 SG level Assumed immediately signal / time de-1 minute 1 minute 0 sec (no delay) lay for AFWS flow d.

SG water level 32% span NA 0% span, 42,030 lbm NA at time of re-54,900 lbm actor trip e.

Initial SG in-111,100 lbm/SG 65,205 lbm/SG 88,830 lbm/ ruptured SG Consistent with power ventory at 544.6 F 78,570 lbm/ intact SG Rate of change See Figure Q410.33-1 NA See Figure Q410.33-2 NA before & after and -3 AFWS actuation Decay heat ANS + 20%

ANS + 20%

ANS + 20%

ANS + 20%

f.

AFW pump design 1225 psia 1225 psia 1225 psia 1225 psia pressure 410-31 Rev. 6 New Page

TABLE Q410.33-2 (Page 2)

SUMMARY

OF ASSUMPTIONS USED IN AFWS DESIGN VERIFICATION ANALYSES Loss of Feedwater Main Steamline Break Transient (Station Blackout)

Cooldown Main Feedline Break (containment) g.

Minimum # of SGs Divided equally among N/A Loop 1 - 0%

NA which must re-2 SG Loop 2 - 0%

ceive AFW flow Loop 3 - 50.0%

Loop 4 - 50.0%

(Loop 1 is the broken loop) h.

RC pump status Tripped at reactor Tripped Operating and trip-All operating trip ped at reactor trip.

i.

Maximum AFW 134 F 134 F 134 F 134*F temperature j.

Operator action None NA SI terminated 30 10 minutes minutes after reactor trip k.

MFW purge volume /

41.5 ft3/445*

150 ft3/440 41.5 ft3/445 F 500 ft3/ loop (for SG and temperature dryout time) 1.

Normal blowdown None assumed None assumed None assumed None assumed m.

Sensible heat See cooldown Table Q410.33 See cooldown NA

-3 i

n.

Time at standby /

2 hr/4 hr 2 hr/4 hr 2 hr/4 hr NA time o.

AFW flow rate 492 gpm - constant Variable 492 gpm - constant 1468 gpm (constant) to broken SG 410-32 Rev. 6 New Page

TABLE Q410.33-3

SUMMARY

OF SENSIBLE HEAT SOURCE Primary Wster Sources (initially at ESD power temperature and inventory)

RCS fluid Pressurizer fluid (liquid and vapor)

Primary Metal Sources (initially at ESD power temperature)

Reactor coolant piping, pumps and reactor vessel Pressurizer Steam generator tube metal and tube sheet Steam generator metal below tube sheet Reactor vessel internals Secondary Water Sources (initially at ESD power temperature and inventory)

Steam generator fluid (liquid and vapor)

Main feedwater purge fluid between steam generator and AFWS piping.

Secondary Metal Sources (initially at ESD power temperature)

All steam generator metal above tube sheet, excluding tubes.

410-33 Rev. 6 New Page

l 880" N

O 0

I 1

0 T -

0 A3 0

T3 l

1 S

0 l

T R1 U A 4 l

0 E l

K L Q C C l

A U e L N r B

u Y

l N

i R

A g B

A O

F W

I I

L TA 1

R AT I

GE X

TA NT U

SC A

A I

VW l

G IED N

in o

CE I

r EE V

t RF I

E SY C

0 0

P R E

O A RW l

0 l

1 OI O

l LL T

L I W O F OX O N

l WUL R

l S E TAF P

)

O T l

S A

D O

LW l

O N

D O E C

WE E

TF l

S

(

E M

I l

T 0

l 0

l 1

l l

l l

l l

l 0

1 0

0 0

0 0

0 O

0 0

0 0

0 O

2 1

0 9

8 7

G 5

4 3

2 1

1 1

1 5o'u.Ov>0 5og e 2 xOQm* mo Egli>

c ll1 ll l

l1 1

lll!f1

l 1

l h3 -

NO I

2 0 H T

0 T A

I 3

0 T

W S 3 0

l 1

l E

0 R

R 1

A 4 U

l E

TR

~

L l

~ PE Q

~

C W

UW

~ ROU e G

~

N r O

l P

NI L E

u VF

~ NEA g l

I TBi GW IE R -

LIWF N O C E DSA IV L E T l

EFT EFA F

RA FOC I

E W

C R T

E E O D l

E R T N

M'.s,M' A

E ta S

PF PW e

O r

Y OD O

ta R

L E A

0 __

OE L

h F

TI 0

T CL 0

CY AI 1

l X

AR T

l U

TA N

NI A

I l

L I I OX l

)

WU TA l

SD N

l O

CE S

(

E M

I l

T 00 l

1 l

l l

l l

l N

l E

KP OO RO l

BL 0

1 0

0 0

0 0

0 0

0 0

0 0

0 0

2 1

0 9

8 7

6 5

4 3

2 1

1 1

1 53 omg 6 oIt s co <E2 c s<fm c

l 1

1ll' l1 1ll l

l)l

1 1llI l

e

$@w T

N U

0 O

O 0

I H

T 0

T A

1 0

I 1

T W

S 1

l E

R 3

R A

U E

l TR L 3

W I

PE C 3 UW J G

RO A 0 O

P 1

P NI L I

E 4

A VF NE B Q I

ITLI W E R I

DS A C E E T EF T EF A RA FO C W

TO D l

rg, t,'

e E

i No N

w GW E

r PF N O xO O

a Y

I n

VL

{'

o R

F 0

I L

E A

0 C R TI i

10 E E CL R T AIX l

A T

S U

l PW N

OD I A I

OE I

S

)

L E F

f D

T N

CY I

O AR C

TA E

NI S

L I I

(

OX E

M WU i

T TA I

00 l

1 l

i l

I

]

I N

E I

KP OO RO BL i

O 0

0 0

0 0

0 0

0 0

0 0

o 2

1 0

9 8

7 6

5 3

2 1

1 1

1

-E:8$58oeysbN

& k;m i

.l I

f

CNS 420.5 On November 7, 1979, Westinghouse notified the Commission of a po-tential undetectable failure which could exist in the engineered safeguards P-4 interlocks.

Test procedures were developed to detect failures which might occur.

The procedures require the use of volt-age measurements at the terminal blocks of the reactor. 2p breaker cabinets.

Concern:

In order to minimize the possibility of accidental shorting or ground-ing of safety system circuits during testing, the staff believes that suitable test jacks should be provided to facilitate testing of the P-4 interlocks.

Provide a discussion on how the above issue will be resolved for Catawba.

Response

In order to implement the Westinghouse recommended procedures, a voltage indicator will be wired to the reactor trip breaker terminal blocks.

This will allow operating personnel to check the status of the P-4 interlock.

This modification will be completed prior to fuel load.

420.6 Safety Injection Pump Suction Isolation Valve NI 100B and Safety Injection Pump Miniflow Header to Feedwater Valve NI 147B require power lockout to meet the single failure criterion.

The power lockout scheme for each valve, as shown on Catawba Drawings CNEE-0151-01.10 and 0151-01.13, uses an additional manually controlled contactor (M2).

Concern:

The staff believes that a short of the #1-#2 contact set for either

" MAINTAINED" switch (NI65 or NI73) would constitute a non-detectable failure and thus violate the single failure criteria.

Provide a discussion of how the above will be resolved for Catawba.

420.7 TMI-2 Action Plan Item II.E.1.2 Part 2 requires safety grade indica-tion in the control room of auxiliary feedwater flow to each steam generator powered from emergency buses consistent with emergency power diversity requirements of Auxiliary Systems Branch Technical Position 10-1.

Concern:

l The applicant's response to this Action Plan Item in the FSAR is inadequate and the staff believes changes are being made in this area.

Provide a discussion of the power sources to be used for the auxiliary feedwater flow indication.

420-9 Rev. 6 New Page i

CNS

Response

TMI-2 Action Plan Item II.E.1.2 Part 2 " Auxiliary Feedwater System Flowrate Indication," Section Changes to Previous Requirements and Guidance states:

The requirements for Westinghouse (W) and Combustion Engineering (C-E) plants have been relaxed to require only a single-channel flow 'ndication, instead of redundant channels.

This single chan-nel need not be seismically qualified nor need it be powered from a Class 1E power source.

The auxiliary feedwater flow indication requirements have been re-laxed for PWRs with U-tube steam generators because flow indication is of secondary importance in assuring steam. generator cooling cap-ability for steam generators of this design.

The auxiliary feedwater flow indication for the Catawba Nuclear Station follows the requirements as set forth in NUREG-0737.

Sin-gle channel monitoring and indication is provided in the control room for each steam generator loop auxiliary feedwater flow.

High reliability battery-backed power sources for the instrumentation are selected in conformance with auxiliary systems branch technical position 10-1.

Failure of one power source will not cause a loss of flow indication to all steam generators.

420-10 Rev. 6 New Page

~. -. ---

CNS 30.7 In Section 8.3.1.1.1.3 of the FSAR you address an automatic transfer (8.3.1.1) scheme at the 6.9 kv level.

Indicate whether the transfer is an auto-matic fast transfer or slow transfer and provide the transfer times involved.

For slow source transfer also indicate the sequence of events involved in the transfer.

Response

See revised Section 8.3.1.1.1.3.

430.8 Provide a guide to symbols and nomenclature for the electrical in-(1.7, 7.0, strumentation and control drawings that are listed in Table 1.7.1-1.

8.0)

Response

See Figure 1.7.1-1.

O 430.9 The undervoltage tripping scheme described in Section 8.3.1.1.2.1 of (8.3.1.1) the FSAR as presently designed is not acceptable:

1.

The setpoint of 83.2% is below the normally specified minimum continuous operating equipment voltage of 90% according to ANSI C84.1.

2Property "ANSI code" (as page type) with input value "ANSI C84.1.</br></br>2" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..

Starting the diesel instantaneously when the voltage drops below 83.2% creates the possibility of unnecessarily challenging the diesel start systems as a result of normal motor starting or~ dis-tribution system short duration voltage transients.

3.

The time delay of 8.5 seconds for any pitage between 0% and 83.2% will likely allow equipment to be damaged since it is possible they could be operated at voltages much lower than their rated values for 8.5 seconds before being separated from offsite power.

A more acceptable undervoltage protection scheme is provided by two levels of undervoltage protection as described in the following re-vised staff position.

The first level will separate the loads rapidly 430-3 Rev. 5

CNS (1 sec) for very low voltage conditions or total loss of voltage, and the second level will allow longer time delays at voltages just below equipment ratings.

The revised criteria, for design of the 2nd level of voltage protection, allows a period of time for the operator to take action to improve the low voltage condition.

This is a pre-ferred method over the previous NRC criteria but in no way detracts from the acceptability of designs in accordance with the previous criteria.

In lieu of justifying the deficiencies in your existing design itemized above you may opt for the revised design criteria contained in part 1 following; however the requirements for blocking the load shedding feature, for optimizing voltage levels at safety-related buses, and for a verification test which are contained in parts 2, 3 and 4 following should be adhered to:

1.

In addition to the undervoltage scheme provided to detect loss of offsite power at the Class 1E buses, a second level of under-voltage protection with time delay should also be provided to protect the Class 1E equipment; this second level of under-voltage protection shall satisfy the following criteria:

a)

The selection of undervoltage and time delay setpoints shall be determined from an analysis of the voltage re-quirements of the Class IE loads at all onsite systems distribution levels; b)

Two separate time delays shall be selected for the second level of undervoltage protection based on the following conditions:

1)

The first time delay should be of a duration that establishes the existence of a sustained degraded voltage condition (i.e., something longer than a motor starting transient).

Following this delay, an alarm in the control room should alert the operator to the degraded condition.

The subsequent occurrence of a safety injection actuation signal (SIAS) should immediately separate the Class 1E distribution system from the offsite power system.

2)

The second time delay should be of a limited duration such that the permanently connected Class 1E loads will not be damaged.

Following this delay, if the operator i

has failed to restore adequate voltages, the Class 1E distribution system should be automatically separated from the offsite power system.

Bases and justification must be provided in support of the actual delay chosen.

c)

The voltage sensors shall be designed to satisfy the fol-lowing applicable requirements derived from IEEE Stan.t*-d 279-1971, " Criteria fcr Protection Systems for Nuclear Power Generating Stations:"

430-4 Rev. 3 New Page

C)$

(m'v')

1)

Class IE equipment shall be utilized and shall be physically located at and electrically connected to the Class IE switchgear.

2)

An independent scheme shall be provided for each divi-sion of the Class IE power system.

3)

The undervoltage protection shall include coincidence logic on a per bus basis to preclude spurious trips of the offsite power source; 4)

The voltage sensors shall automatically initiate the disconnection of offsite power sources whenever the voltcqe set points and time delay limits (cited in item 1.b.2 above) have been exceeded; 5)

Capability for test and calibration during power opera-tion shall be provided.

6)

Annunciation must be provided in the control room for any bypasses incorporated in the design.

d)

The Technical Specifications shall include limiting condi-tions for operations, surveillance requirements, trip set-points with minimum and maximum 1imits, and alIowable values for the second-level voltage protection sensors and associated time delay devices.

2.

The Class IE bus load shedding scheme should automatically pre-vent shedding during sequencing of the emergency loads to the bus.

The load shedding feature should, however, be reinstated upon completion of the load sequencing action.

The technical specifications must include a test requirement to demonstrate the operability of the automatic bypass and reinstatement fea-tures at least once per 18 months during shutdown.

In the event an adequate basis can be provided for retaining the load shed feature during the above transient conditions, the setpoint value in the Technical Specifications for the first level of undervoltage protection (loss of offsite power) must specify a value having maximum and minimum limits.

The basis for the setpoints and limits selected must be documented.

3.

The voltage levels at the saiety-related buses should be optimized for the maximum and minimum load conditions that are expected throughout the anticipated range of voltage variations of the offsite power sources by appropriate adjustment of the voltage tap settings of the intervening transformers.

fhe tap settings selected should be based on an analysis of the voltage at the terminals of the Class 1E loads.

The analyses performed to O

430-5 Rev. 5

CNS O

determine minimum operating voltages should typically consider maximum unit steady state and transient loads for events such as a unit trip, loss of coolant accident, startup or shutdown; with the offsite power supply (grid) at minimum anticipated voltage and only the offsite source being considered available.

Maximum voltages should be analyzed with the offsite power supply (grid) at maximum expected voltage concurrent with minimum unit loads (e.g. cold shutdown, refueling).

A separate set of the above analyses should be performed for each available connection to the offsite power supply and the results forwarded to the NRC.

4.

The analytical techniques and assumptions used in the voltage analyses cited in item 3 above must be verified by actual measure-ment.

The verification and test should be performed prior to initial full power reactor operation on all sources of offsite power by:

a) loading the station distribution buses, including all Class 1E buses down to the 120/208 volt level, to at least 30%;

b) recording the existing grid and Class 1E bus voltages and bus loading down to the 120/208 volt level at steady state conditions and during the starting of both a large Class 1E and non-Class IE motor (not concurrently);

NOTE:

To minimize the number of instrumented locations, (recorders) during the motor starting transient tests, the bus voltages and loading need only be recorded on that string of buses which previously showed the lowest analyzed voltages from item 3 above.

c) using the analytical techniques and assumptions of the previous voltage analyses cited in item 3 above, and the measured existing grid voltage and bus loading conditions recorded during conduct of the test, calculate a new set of voltages for all the Class IE buses down to the 120/208 volt level; d) compare the analvtically derived voltage values against the test results.

i With good correlation between the analytical results and the test results, the test verification requirement will be met.

That is, the validity of the mathematical model used in perfor-mance of the analyses of item 3 will have been established; therefore, the validity of the results of the analyses is also established.

In general the test results should not be more than 3% lower than the analytical results; however, the dif-ference between the two when subtracted from the voltage levels 430-6 Rev. 3 New Page

CNS determined in the original analyses should never be less than the Class IE equipment rated voltages.

Response

A second level undervoltage scheme will be incorporated in the design.

The design of this scheme is under review at the present.

Information will be provided upon design completion.

430.10 Recent experience with Nuclear Power Plant Class IE electrical sys-(8.3.1) tem equipment protective relay applications has established that re-lay trip setpoint drifts with conventional type relays have resulted in premature trips of redundant safety related system pump motors when.the safety system was required to be operative.

While the basic need for proper protection for feeders / equipment against permanent faults is recognized, it is the staff's position that total non-availability of redundant safety systems due to spurious trips in protective relays is not acceptable.

Provide a description of your circuit protection criteria for safety systems / equipment to avoid incorrect initial setpoint selection and the above cited protective relay trip setpoint drift problems.

Response

A description of the circuit protection criteria for safety systems /

equipment to avoid incorrect initial setpoint selection and protec-tive relay trip setpoint drift problems is described in Sections 8.3.1.1.2.1, 8.3.1.1.2.2, and 8.3.1.1.3.4.

430.11 Provide a listing of the following for the containment electrical (8.3.1) penetrations by voltage Class:

I2t ratings, maximum predicted fault currents, identification of maximizing faults, protective equipment setpoints, and expected clearing times.

Provide a description of the physical arrangement utilized in your design to connect the field cables inside containment to the contain-ment penetration, e.g. connectors, splices, or terminal blocks.

Pro-vide supportive documentation that these physical interfaces are qualified to withstand a LOCA or steam line break environment.

Response

Refer to Table Q430.11-1 and Section 8.3.1.4.5.2.

P 430-7 Rev. 6

CNS 430.18 Explicitly identify all non-Class 1E electrical loads which are or (8.3) may be powered from the Class 1E a-c and d-c systems.

Also, for each load identified, provide the horsepower or kilowatt rating for that load and identify the corresponding bus number from which the load is powered.

Response

See revised Section 8.3.1.1.2.2.

Table 8.3.1-1, Sheet 3, has been revised to show that the AC emergency lighting panelboard (non-Class 1E) would be disconnected if an accident signal were initiated.

430.19 In Section 8.3.1.1.3.4 of the FSAR you state that the setpoint of (8.3.1.1) the diesel generator overspeed trip is above the maximum engine speed on a full-load rejection.

Provide the full load engine speed and maximum safe engine speed.

In accordance with position C.4 of Regulatory Guide 1.9 verify that, during recovery from transients caused by step load increases or resulting from the disconnection of the largest single load, the speed of the diesel generator unit does not exceed the nominal speed plus 75 percent of the difference between nominal speed and the overspeed trip setpoint or 115 percent of nominal, whichever is lower.

Response

See revised Section 8.3.1.1.3.4.

430.20 Section 8.3.1.1.3.5 of the FSAR states that the load shedding feature (8.3.1.1) for the Class IE buses will remain blocked following load sequencing until the load sequencer is manually reset or diesel engine speed decreases below approximately 44%.

Branch Technical Position PSB-1 in the Standard Review Plan (NUREG 0-800) requires automatic rein-statement of the load shedding feature upon completion of load se-quencing.

Your present design is not acceptable since it will not automatically result in load shedding upon trip of the diesel gen-erator circuit breaker when there is no loss of diesel generator frequency.

Response

See revised Section 8.3.1.1.3.5.

430-13 Rev. 6 L

CNS The High level setpoint is established sufficiently below the FWST overflow to account for instrument error (approximately 11,490 gallons) and provide warning of imminent overflow.

The Makeup level setpoint provides a volume below overflow level to conservatively account for thermal contraction level variation and minor uses (approximately 6,000 gallons) plus an allowance for instrument error (approximately 11,490 gallons).

This setpoint is approximately 25,000 gallons more than the proposed Technical Spec-ification requirement of 350,000 gallons, which provides an adequate

" working allowance."

The minimum time from the initiation of a LOCA to the start of switchover to recirculation is the minimum time required to reduce FWST volume from Makeup level to the Low level.

This will always ex-ceed 11 minutes with maximum FWST outflow (reference Section 9.2.7.2.e) and most adverse level instrument errors.

The switchover procedure and level setpoints conservatively account for all relevant factors of FWST sizing while maximizing the minimum volume available for cold leg injection, placing appropriate emphasis on timely completion of manual actions of the switchover procedure, and maximizing the containment sump volume.

The FWST alarms are derived from safety grade instrument loops.

440.43 Identify single failures and operator errors that would divert ECCS (6.3, 15.6.5) flow.

For both large and small breaks discuss the effect of these failures on flow to the core, the containment sump water level, and conformance with the 50.46 acceptance criteria.

Response

In both large and small break LOCA analyses for Catawba loss-of-offsite power coincident with the accident is assumed.

The single failure subsequently considered is the loss of a diesel generator so that i

only one train of ECCS flow of the two actually present is considered to be available.

Therefore, for both large and small break LOCAs ECCS flow to the core is at a conservatively low value following its automatic actuation, especially since all water delivered to the broken loop is considered to spill directly to the containment sump.

Notwithstanding these conservatisms, conformance with the 10 CFR 50.46 acceptance criteria is demonstrated in the large and small break LOCA analyses.

No other postulated single failure would have as great an effect on ECCS flow delivery.

The ECCS termination and reinitiation criteria provided in the Catawba Emergency Operating Procedures (EOPs) are designed to minimize any possibility of an operator error improperly or prematurely shuting off safety injection.

Termination criteria for high pressure safety in-jection flow (HPI) following a LOCA event call for a shutoff of all 440-43 Rev. 6 L

CNS HPI when the RCS pressure reaches 2000 psia and is rising, the pres-surizer is more than half full, and steam generators are being fed auxiliary feedwater.

For a break as small as a 0.5 " equivalent" diameter hole, as soon as the HPI is terminated, a rapid depressuriza-tion of the system occurs.

As some steam forms in the RCS, the de-pressurization rate slows, but under the Westinghouse HPI termination guidelines, the operator reinitiates HPI after the RCS depressurizes to the safety injection set pressure.

It is evident that even for this small break, the required pressure drop would occur rapidly.

As the operator continually follows the Westinghouse criteria, and utilizes only the conditions of no HPI at all, or full maximum HPI, the system will oscillate in pressure between 1800 and 2000 psia, but the core will remain covered and adequately cooled.

440.44 Page 6.3-7 states "As the reactor coolant pressure drops below the (6.3, 15.0) nominal pressure in the gas accumulator (1000 psia) the membrane in the gas crossover line will tear open at a differential pressure of about 40 psi and flow will be initiated through the normally open isolation valves and injection lines into the reactor vessel." This is not consistent with some of the Chapter 15 analyses which assume a pressure of about 1250 psia.

For each of the Chapter 15 analyses that take credit for UHI, show that the analysis is based on the correct UHI pressure with appropriate allowances for the minimum accumulator initiation pressures.

440-43a Rev. 6 Carry Over

CNS 440.94 Describe the recovery from the inadvertent opening of a pressurizer (15.6.1) safety or relief valve accident.

Include information on operator action, the pressurizer water level, the potential for void for-mation in the RCS, and actuation of the engineered safety features.

Response

The case of inadvertent opening of a pressurizer safety or relief valve was analyzed generically in WCAP-9600 " Report on Small Break Accidents in W NSSS," Section 3.0.

Two cases of break size were analyzed representing one small PORV and three large PORV's opening, and then sticking in the full open position.

This break size range covers a safety valve opening as well.

The characteristics of both cases were similar as shown in WCAP-9600.

In both cases pressurizer water level rises, Reactor Coolant System (RCS) depressurization occurs resulting in automatic actuation of reactor trip and safety injection (SI) based on low pressurizer pressure signals.

The RCS depressurizes to the point where leak flow equals the SI flow.

If only minimum safeguards safety injection is available, there is voiding in the core and hot legs, but no core uncovery.

The clad temperature remains below steady state operating temperatures, and decay heat is removed via natural circulation.

If maximum safe-guards safety injection is available, depending on the leak size, the RCS may repressurize and return to subcooled conditions.

The scenario of a stuck open pressurizer PORV or safety valve does not represent the limiting small break scenario.

The small cold leg breaks analyzed in the FSAR are the worst small breaks.

1 440-86g Rev. 6

CNS 440.96 Provide an analysis of the transients resulting from a break in the (15.6.5)

ECCS injection lines.

Describe the flow splitting which will occur in the event of the most limiting single failure and verify that the amount of the flow actually reaching the core is consistent with the assumptions used in the analysis.

Show that the 10 CFR 50.46 accep-tance criteria are satisfied.

Response

Any break postulated to occur in the ECCS line is bounded by the spectrum of breaks presented in Section 15.6.

Standard assumptions used in defining safety injection flow for either a large or small cold legs break LOCA analysis include:

1.

The spilling of the broken loop accumulator directly to contain-ment.

2.

The spilling of the safety injection line attached to the broken cold leg.

3.

A single failure condition such that only one train of safety injection pumps operates.

Each of the four RCS cold legs has an injection line attached.

Flow delivered into the RCS is computed based on the following logic:

One train of ECCS pumps starts and delivers flow into the Reactor Coolant System through three branch injection lines.

One branch injection line spills to containment backpressure.

The branch injection line with minimum system resistance is se-lected to spill to minimize reactor delivery.

The flow delivered into the reactor through the reactor coolant pump seals is assumed to be lost and therefore seal injection is not included in the total core delivery.

Safety injection flows computed via this methodology are conser-vatively low for any postulated break location.

440.97 Identify and provide the basis for the most limiting single active (15.6.5) failure for large and small breaks.

Discuss how the accident analy-sis is representative of these failures.

Response

The limiting single active failure for both large and small break LOCA analyses is the loss of a diesel generator.

As a consequence of this failure only one train of safety injection pumps is avail-able, and a minimum amount of pumped injection is delivered to the 440-88 Rev. 6 e

ya I..-

', 7 =#

< q,

,s/

,4 m

.n u t,

s.,.

y

~

')

3 9 y"i CN$

s

Response

i/

'j f,

fj l

r IntheCatawbalargebreak/LOCAanalysisa;coldileg/accumulatorwater

's r

volume of 1050 cubic feet per tank and a nitrogen / gas pressure of 400 psia were assumed.

The Catawba' cold leg acc6Tolator; tank set-points will be Ast,ablished to meet these requirements,'a minimum

~

possible tank.'wat'er volume of 1088 cubic feet and a minitru'm gas pres-sure of 400 p'sia.

The use of the 1050 cubic feet cold lator wat"er volume and a gas pressure value of 400 psla.leiraccumu-

~

in the Ap-

.pendix K large break ECCS performance afalysis corpespends, exactly with the McGuire 'LOCA analysis and baf previoutiy)Sen discussed N-M and justified as appropriate for the very same" plant accculator j

setpoints in McGuire Question 212.105.

The concTusion ' drawn there-in that the calculated peak clad temjerature 0P a UHF philt DECLG.

perfect mixing case is unaffected by assuming {c:aximum and !do.imu

,o M;-

accumulator water delivery rate bounding conditionsipplies<also e

to the Catawba limiting case,(DECLG.C = 1.0).

The.*.0CA'an11yi.is y '3Q'e' D

accumulator values and the actual'.plhnt setpoints are the sIm Catawba and McGuire, as noted above.

The blowdown' transients for T"

the limiting case breaks for the two plants are very similar in 5

'h'

' I predicted accumulator performance and in bypass deff,cate Ee'habior!, V sotheimpactofassumingboundingaccumulatorwaterde' liv,ey/~MesJ is virtually the same.

" S

,.g-+

' ?'

a g/

g "l

e c

j 440.104 Your response to Question 440.12 is not acceptable. ' You. state (5.4.7)

"the residual heat removal flow rate is throttled to abc'ut' 3500 S (440.12) gpm through each of the residual heat removal loops." This' implies s both loops are operating, therefore, your postulated air entrainmente j*

s f) event may preclude the continued use of both operating trains in-5 stead of one train.

Accordingly, provide the information requested

~

in Items 2 and 3 of Question 440.12.

Describe your procedures for RHR operation when the steam generator tubes are drained.

e The staff 4 position is that each train of the RHR will be provided -

,n w,

with an alarm in the control room to alert the operator to RHR de-gradation.

Provide your basis for the alarm setpoint.

y

Response

i w._.,

Having_both RHR trains',in operation during Mode 5 activities, though not! precluded, is un?,tkely.

The heat load is such that one RHR loop

<v

~

cart adequately provide all; of the required cooling.

f The Technical Specifi. cation: (Section 3/4.4.1.4) specifically address-l 3

er ';

PCS loop, drain operations,and specify that both RHR trains be op/

able prior to draini,ng the steam generators.

. f j..

l s

e Alarms are provided in the control room to alert the operator should the unlikely situation of RHR degradation occur.

I l

/

j'

<l 440-92 Rev. '6 l

[

- q -

~w.

m. a.

-w.

/

m t

3 1

l CNS

+

t ih 440.107 Your response to Question 440.28 is not complete.

Provide an evalua-(6.3) tion of,your conformance to Branch Technical Position RSB 6-1, Item

((440.28)

B.5.

10entify and justify any deviations from this position.

I, " ~

q -

~

Response

3 y/

i L.

440.108

,Your response to Question 440.29 is inadequate and is most likely (6.3)

' based on yaur withdrawn FSAR and not the current FSAR.

Provide the (440.23) response to Question 440.29.

Sections 6.3.2.1 and 7 and Table e

/

6.3.2-3 should be consistent and complete.

r r

~

s Responaev t

/

1Ee~responsetoQuestion440.30isincomplete.

Feedwater pipe breaks

~440.109

~..

' (6.3'&

should also.be discussed.

For each type of pipe break in the primary 15.0)-

and secondary systems, provide the information requested in Question (440.29) 440.20.

Time response for operation reaction (credit only given from time of receipt of control room alarm from safety grade instrumenta-tion) should be discussed, and may be based on ANSI N660 criteria when determining accident consequences.

The accident description and discussions of consequences should take into consideration the available mitigating equipment as a function of pressure.

p'. Response:

's

- j

<If a feedline rupture occurs while both SI actuation signals are

._blocked, a low-steam generator water level alarm will be generated followed by a low-low steam generator water level signal.

Auxiliary feedwater flow is initiated on receipt of low-low steam genernor

/

water level signal.

Steamline isolation will occur on high negative i

' steam pressure rate.

An alarm for steamline isolation will alert O

e the operator of the accident.

t

/

440.110 The response to Question 440.36 did not adequately consider the effects (6.3 &

of single failures that could lead to a failure of both safety injec-15.0).

tion pumps.

The staff position regarding these single failures is (440.36) presented in Question 440.106.

t' 440-94 Rev. 5 New Page

,m i

CNS The response to Question 440.36 states "miniflow recirculation paths are also provided for the charging pumps.

These paths are isolated upon receipt of an "S" signal.

The pump deadheading problem is a valid concern for the charging pumps as operating these pumps at or above their shutoff head would lead to failure of the pumps due to overheating.

Analyses have been performed to show that adequate core cooling is provided by flow from the safety injection and re-sidual heat removal pumps."

Identify those conditions that would lead to failure of the charging pumps.

Include the postulated case of a small break in which the RCS pressure remains at or near the RCS safety valve setpoint.

Describe the analysis that was performed to show that adequate core cooling is provided by the safety injection and residual heat removal pumps.

Show that the Chapter 15 analysis has accounted for the potential failure of the charging pumps.

Response

The secenario leading to deadheading of the charging pumps was des-cribed in extended detail in a Westinghouse letter to the Commission:

T. M. Adnerson, letter to V. Stello, subject:

" Centrifugal Charging Pump Operation Following Secondary Side High Energy Line Rupture,"

dated May 8, 1980.

In order to prevent this scenario the charging pumps minimum flow isolation valves no longer close on an "S" signal.

The minimum flow path can be isolated by the control room operator when it is certain that reactor coolant pressure is low enough to assure pump minimum flow requirements.

440.111 Your response to Question 440.37 is not adequate.

You state "In ad-(15.0) dition, the head of water provided by the RWST further ensures the (440.37) lines will remain full and water hammer concerns will not develcp."

This does not appear to be correct.

For example, the charging pump subsystem is not aligned to the RWST during the ECCS standby con-dition.

A small leak in this subsystem could void the charging lines.

Describe the means for ensuring water hammer will not occur in the

~

charging pump lines incluoing specific Technical Specifications test-ing requirements and frequency of tests.

Response

The ECCS lines which incorporate the charging pumps consist of:

1.

RWST to charging pump suction isolation valves 1NV252A and 1NV2538 -

The water head in the RWST precludes voiding in these lines due to a small leak.

2.

Suction isolation valves to common suction line check valve 1NV254 -

440-95 Rev. 6

CNS In order for us to complete our analysis, we request that you pro-vide the length of operation of the ice condenser (including the start and end times) and the delay time before startup of the recir-culation fans.

Response

The time dependent ice condenser removal efficiency is presented as a comparison to demonstrate some of the conservatism in the LOCA analysis.

Assumption 9 in Section 15.6.5.4.1 describes the ice con-denser elemental iodine removal efficiency actually used in the LOCA dose analysis.

The length of operation of the ice condenser is given in Section 15.6.5.4.1, Assumption 9.

The length of operation includes a 10 minute delay time before the startup of the containment air return fans.

450.03 For the accident of rod ejection with loss of offsite power, the (15.4.8) following parameters are needed for our analysis:

(1) The time required for the pressures of primary side and sec-ondary side to equalize.

(2) The time required to start shutdown cooling.

Response

(1) The primary and secondary pressures essentially equilibrate at 3000 sec after break initiation.

(2) Switchover to RHR cooling would be contingent on unit recovery and response following a small break LOCA.

The Westinghouse Owners Group has developed Emergency Response Guidelines (ERG's) in response to NUREG-0737, item I.C.1 which contain instructions for a post LOCA cooldown and depressurization.

These ERG's will be used in development of the Catawba Emergency Procedures.

A 450.04 The radiological consequence analysis for the steam generator tube (15.6.3) rupture accident assumes that the affected steam generator can be isolated in 30 minutes.

Based on Figures 15.6.3-2 and 15.6.3-4 it seems very unlikely that the affected steam generator could be iso-lated.

Therefore, provide additional information which supports the assumption that the affected steam generator can be isolated within 9

450-2 Rev. 5 m

CNS 30 minutes or provide additional steam release volumes from the af-fected steam generator resulting from the steam dump until the af-fected steam generator can be isolated.

Also provide both the amount of primary to secondary leakage in the affected steam generator prior to the reactor trip and the amount of primary to secondary leakage after reactor trip up to the time of steam generator isolation.

Response

The analysis assumptions are conservatively designed to maximize doses and do not explicity model operator actions necessary for recovery.

Since these actions, which are described in Section 15.6.3.2, are not modeled, primary pressure, Figure 15.6.3-2, is predicted to remain greater than the faulted steam generator pres-sure, Figure 15.6.3-4, for the duration of the analysi-Sufficient controls and instrumentation are available to complete the necessary recovery actions from within the control room.

Hence, 30 minutes is considered adequate time for a design basis tube rupture to ter-minate releases from the faulted steam generator.

Additional time would be available for smaller tube failure events since primary-to-secondary leakage would be less.

The analysis assumptions leads to a conservative estimate of 25,700 lbm of reactor coolant transferred to the faulted steam generator prior to reactor trip.

An additional 76,600 lbm primary-to-secondary carryover is estimated to occur until isolation is complete.

450.5 In the analysis of the offsite radiological consequences of the (15.6.5)

Design Basis Accident (FSAR Section 15.6.5.4.1), provide the fol-lowing information:

(1) operating characteristics of the Annulus Ventilation System following a LOCA, including iodine filter efficiencies, total system flow, recirculation flow, pressurization flow, time at which total, pressurization and recirculation flows are initiated, description of time dependent characteristics of recirculation and pressurization flows, and a schematic showing leakage paths, flow rates, filter efficiencies.

(2) free volumes of the upper compartment, lower compartment, and annulus.

Response

The requested information is presented in Table 6.2.3-2, Table 15.6.5-10, and Figure Q450.05-1.

450-3 Rev. 6 L

i CNS

Response

1 A response will be provided in a future revision.

480.3 The review of this section cannot be completed until after the infor-(6.2.3.2) mation identified as "later" in Table 6.2.3-3 is submitted.

Either (AR) provide this information or provide a schedule for submittal of the information.

Response

See revised Table 6.2.3-3.

480.4 Discuss the design features which prevent the release of fluids from (6.2.1.1.2) high energy lines into the annulus between the primary and secondary containments or provide an analysis to demonstrate the ability of j

the containment to withstand the effects of rupture of the largest i

high energy line within the annulus.

Response

All high energy penetrations consist of the " Hot Penetration" assembly as described in Section 3.6.2.4.

e 480.5 FSAR Table 6.2.1-4 describes the structural heat sinks used in the (6.2.1.1.3.1) analysis of long-term containment pressure response to LOCAs.

Identify Slabs 1 through 14, in the same manner as was done for Slabs 15 through 20.

Response

See revised Table 6.2.1-4.

i 480.6 Provide an analysis verifying that containment pressure will be re-l (6.2.1.1.3.1) duced, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the beginning of the accident, to no more than one half of the peak pressure experienced during the DBA LOCA.

i

Response

Figure Q480.6-1 shows the extended pressure transient out to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Under NUREG-0800, 6.2.1.1 B, for ice condenser containments,Section II, Item 3 states pertaining to "G.D.C.

38 as it relates to the con-tainment heat removal system (s) function to rapidly reduce the contain-ment pressure and temperature following any loss-of-coolant accident."

Figures Q480.6-1, Q480.9-1, and Q480.9-2 show that to be true.

The l

reduction of the containment pressure to less than 50% of the peak calculated pressure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is a criteria for acceptance for a dry containment (Section 6.2.1.1 A of NUREG-0C00).

480-3 Rev. 6

~~

1631 1 14.0 12.5 10.0 8ma 7.5 g

8 e

!E 5.0 2.5 i

I II I

I II I

I II I

I II o

1 2

4 5

10

'10 103 10 10 TIME (SECONDS) l CONTAINMENT PRESSURE TRANSIENT ova rca CATAWBA NUCLEAR STATION Figure Q480.6-1 l

[

CNS 480.7 For the long-term LOCA containment pressure response analysis (using (6.2.1.1.3.1)

LOTIC-1), specify the temperature assumed for the Refueling Water Storage Tank water.

Response

The Refueling Water Storage Tank water temperature used was 114*F.

480.8 In FSAR Section 6.2.1.1.3.1, the analysis of maximum reverse differ-(6.2.1.1.3.1) ential pressure across the operating deck (using LOTIC-2) gave a result of 0.65 psi.

Specify the design reverse differential pressure capability of the operating deck, steam generators and pressurizer enclosures, and ice condenser lower inlet doors.

Verify that an adequate margin is provided between design and maximum calculated reverse differential pressure.

Response

The maximum reverse differential pressure of 0.65 psi is the result of a very conservative analysis as described in Section 6.2.1.1.3.1.

This pressure is fairly small in comparison to other loads which act on pressure barrier structures (operating deck, pressurizer en-closure etc.).

While this pressure differential was not included in the design of these structures, the conservative nature of their design does allow for significant reverse differential pressure to occur.

For example, the pressurizer enclosure ccncrete structure uses the same reinforcement in both faces of the concrete.

This ef-fectively allows for significant pressure differentials to occur in either direction.

The operating deck is designed for a live load of 300 lb/ft2 in either direction.

Similar conservative assumptions were used in the analysis and design of all presssure barrier structures.

In conclusion, the magnitude of other loads and the conservative manner in which these loads were applied make the reverse differen-tial pressure of 0.65 psi relatively insignificant.

480.9 In order to review the environmental qualification of equipment (6.2.1.1.3.1, inside containment that is required to operate during an accident, 6.2.1.1.3.2.2) the containment atmoshpere temperature and pressure transients for the postulated accidents (LOCAs and MSLBs) must be known.

There-fore, extend the curves shown in FSAR Figures 6.2.1-5, 6.2.1-6, 6.2.1-15, 6.2.1-16, and 6.2.1-17 to display these data for the time from the beginning of the accident until equilibrium conditions are reached in the containment atmoshpere.

Provide figures showing the containment pressure transients corresponding to the temperature transients shown in Figures 6.2.1-15, 6.2.1-16, and 6.2.1-17, over the same time mentioned above.

Also, in Section 6.2.1.1.3.2.2, on 480-4 Rev. 6

CNS page 6.2-18 of the FSAR, a " Figure C" is referenced, but is appar-ently not included in the FSAR.

Provide this figure, and the cor-i responding pressure transient, displaying the information over the same time range as discussed above.

j

Response

Figures Q480.9-1 and -2 show the containment transients out to 24 j

hours.

1 1

l i

i i.

4 i

[

1 i

i l

l 480-4a Rev. 6 I

Carry Over i

~.., _

1 G31 -2 260 250 i

225 200 C

9 W

W 5 175

<C b

g N

150 125 i

l 100 I

I ll l

l ll 1

I II I

III 80 101 102 3

10 104 5

10 l

TIME (SECONDS) t

~

CONTAINMENT UPPER COMPARTMENT TEMPERATURE TRANSIENT out powie CATAWBA NUCLEAR STATION W'

Figure Q480.9-1 l

1631-3 260 250 -

225 200 IC O

w E

h 175 5

m E

I" 150 125 100 80 1

2 4

5 10 10 1g3 1g 10 TIME (SECONDS)

CONTAINMENT LOWER COMPARTMENT OTEMPERATURETRANSIENT CATAWBA PAJCLEAR STATION Figure Q480.9-2

g CNS 480.10 For the analysis of steam line breaks using the LOTIC-3 code, the (6.2.1.1.3.2.2) assumption of complete revaporization of condensate for large breaks was used.

It is our position that no more than 8% revaporization should be assumed for steam line break analyses.

Therefore, revise the analysis using an assumption of no more than 8% revaporization.

Response

For large steamline breaks inside of an ice condenser containment the model presented in Reference 1 was used.

In large steamline breaks entrainment carryover is expected during the blowauwa, and the high degree of turbulence makes revaporization of the condensate a realistic assumption.

In addition, the small steamline breaks are the most limiting breaks, where no condensate revaporization is assumed and a convective heat flux model is utilized as described in Reference 1.

The small steamline break model described above yields similar results to the NRC recommended model assuming 8%

revaporization.

References:

1.

Hsieh, T. and Liparulo, N.

J., " Westinghouse Long Term Ice Con-denser Code - LOTIC-3 Code," WCAP-8354-P Sup. 2, February 1979.

480.11 Provide the results of sensitivity studies which demonstrate that (6.2.1.1.3.2.2) the values of containment temperature, pressure and relative humidity assumed as initial conditions represent conservative values for the containment temperature response analysis of the assumed main steam line break.

Response

Use of high initial containment temperatures will yield a lower wall heat removal rate.

Therefore, high initial temperatures are conservative for determining the containment temperature response.

The containment MSLB temperature transients are insensitive to initial conditions assumed for the pressure and relative humidity.

Changes to both the pressure and relative humidity effect the partial pressures of the steam and air.

A 2 psi change in the total pressure would change the partial pressure of steam by approximately.2 psi, this would change the steam enthalpy by about.1%.

The air, which acts like an ideal gas, is not affected at all by changes in its partial pressure.

The total change in enthalpy would then be much less than the.1% because the initial air mass is more than 10 times the initial steam mass.

Changes in relative humidity that would re-sult in a.2 psi change in the steam partial pressure would effect the enthalpy in a similar way.

480-5 Rev. 6

-